Compact nuclear reactor with integral steam generator

ABSTRACT

In an illustrative embodiment, a pressurized water nuclear reactor (PWR) includes a pressure vessel ( 12, 14, 16 ), a nuclear reactor core ( 10 ) disposed in the pressure vessel, and a vertically oriented hollow central riser ( 36 ) disposed above the nuclear reactor core inside the pressure vessel. A once-through steam generator (OTSG) ( 30 ) disposed in the pressure vessel includes vertical tubes ( 32 ) arranged in an annular volume defined by the central riser and the pressure vessel. The OTSG further includes a fluid flow volume surrounding the vertical tubes and having a feedwater inlet ( 50 ) and a steam outlet ( 52 ). The PWR has an operating state in which feedwater injected into the fluid flow volume at the feedwater inlet is converted to steam by heat emanating from primary coolant flowing inside the tubes of the OTSG, and the steam is discharged from the fluid flow volume at the steam outlet.

CLAIM OF PRIORITY

This application is a continuation of U.S. patent application Ser. No.12/891,317, filed Sep. 27, 2010, now U.S. Pat. No. 9,343,187, the entiredisclosure of which is incorporated by reference herein.

BACKGROUND

The following relates to the nuclear reactor arts, steam generator andsteam generation arts, electrical power generation arts, and relatedarts.

Compact nuclear reactors are known for maritime and land-based powergeneration applications and for other applications. In some such nuclearreactors, an integral steam generator is located inside the reactorpressure vessel, which has advantages such as compactness, reducedlikelihood of a severe loss of coolant accident (LOCA) event due to thereduced number and/or size of pressure vessel penetrations, retention ofthe radioactive primary coolant entirely within the reactor pressurevessel, and so forth.

Disclosed herein are further improvements that provide reduced cost,simplified manufacturing, and other benefits that will become apparentto the skilled artisan upon reading the following.

BRIEF SUMMARY

In one aspect of the disclosure, an apparatus comprises: a generallycylindrical pressure vessel defining a cylinder axis; a nuclear reactorcore disposed in the generally cylindrical pressure vessel; a centralriser disposed coaxially inside the generally cylindrical pressurevessel, the central riser being hollow and having a bottom end proximateto the nuclear reactor core to receive primary coolant heated by thenuclear reactor core, the central riser having a top end distal from thenuclear reactor core; and a once-through steam generator (OTSG)comprising tubes arranged parallel with the cylinder axis in an annularvolume defined between the central riser and the generally cylindricalpressure vessel, primary coolant discharged from the top end of thecentral riser flowing inside the tubes toward the nuclear reactor core,the OTSG further including a fluid flow volume having a feedwater inletand a steam outlet wherein fluid injected into the fluid flow volume atthe feedwater inlet and discharged from the fluid flow volume at thesteam outlet flows outside the tubes in a direction generally oppositeflow of primary coolant inside the tubes.

In another aspect of the disclosure, an apparatus comprises: apressurized water nuclear reactor (PWR) including a pressure vessel, anuclear reactor core disposed in the pressure vessel, and a verticallyoriented hollow central riser disposed above the nuclear reactor coreinside the pressure vessel; and a once-through steam generator (OTSG)disposed in the pressure vessel of the PWR, the OTSG including verticaltubes arranged in at least one of (i) the central riser and (ii) anannular volume defined by the central riser and the pressure vessel, theOTSG further including a fluid flow volume surrounding the verticaltubes; wherein the PWR has an operating state in which feedwaterinjected into the fluid flow volume at a feedwater inlet is converted tosteam by heat emanating from primary coolant flowing inside the tubes ofthe OTSG, and the steam is discharged from the fluid flow volume at asteam outlet.

In another aspect of the disclosure, a method comprises: constructing aonce-through steam generator (OTSG), the constructing including mountingtubes of the OTSG under axial tension; and operating the OTSG at anelevated temperature at which the tubes are under axial compression.

BRIEF DESCRIPTION OF THE DRAWINGS

The invention may take form in various components and arrangements ofcomponents, and in various process operations and arrangements ofprocess operations. The drawings are only for purposes of illustratingpreferred embodiments and are not to be construed as limiting theinvention.

FIG. 1 diagrammatically shows a perspective partial-sectional view anuclear reactor including an integral steam generator as disclosedherein.

FIG. 2 diagrammatically shows a side sectional view of the upper vesselsection of the nuclear reactor of FIG. 1 with the tubes of the steamgenerator omitted to emphasize the downcomer volume.

FIG. 3 diagrammatically shows Section D-D indicated in FIG. 2.

FIG. 4 diagrammatically shows flow of the primary and secondary coolantfluids in the integral steam generator of FIG. 1.

FIG. 5 diagrammatically shows an illustrative process for manufacturingand deploying the integral steam generator of FIG. 1.

FIG. 6 diagrammatically illustrates the upper pressure vessel portion ofa variant embodiment.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

With reference to FIG. 1, a perspective partial-sectional view anillustrative nuclear reactor is shown. A nuclear reactor core 10 isdisposed inside a generally cylindrical pressure vessel. In theillustrative embodiment the pressure vessel includes a lower pressurevessel portion or section 12 housing the nuclear reactor core 10, anupper vessel portion or section 14, and mid-flange region 16. This ismerely an illustrative configuration, and the pressure vessel can ingeneral be constructed of as few as a single portion or section, or twoportions or sections, three portions or sections (as illustrated), fourportions or sections (for example including a fourth upper “cap” portionor section separate from the upper portion or section), or so forth. Thepressure vessel 12, 14, 16 contains primary coolant, which in theillustrative case of a light water reactor is water (H₂O), optionallyincluding other additives for reactivity control, such as a boroncompound (e.g., “borated water”). In other contemplated embodiments theprimary coolant may be another fluid, such as heavy water (D₂O). Theprimary coolant fills most or all of the volume of the pressure vessel12, 14, 16. A reactor inlet annulus 18 surrounds the reactor core 10 toenable primary coolant to flow to the reactor core 10. Optionalshielding or shrouding 20 disposed in the reactor inlet annulus 18provides additional radiation shielding from the reactor core 10. Theillustrative reactor is a pressurized water reactor (PWR) in which theprimary coolant is sub-cooled light water maintained under an elevatedpressure at a temperature below the boiling point (saturationtemperature) at the operating pressure; however, a boiling water reactor(BWR) in which the primary coolant operates at the saturationtemperature at an elevated pressure, or another reactor configurationsuch as a configuration employing heavy water, is also contemplated.

Reactor control is provided by upper internal neutron-absorbing controlrods 22 and a control rod drive mechanism (CRDM) 24 that is configuredto controllably insert and withdraw the control rods into and out of thenuclear reactor core 10. Diagrammatic FIG. 1 only identifies twoillustrative control rods 22; however, in some embodiments the controlrods may number in the dozens or hundreds, with insertion pointsspatially distributed across the reactor core area to collectivelyprovide uniform reaction control. The CRDM 24 may be divided intomultiple units (details not illustrated), each controlling one or morecontrol rods. For example, a plurality of control rods may beoperatively coupled with a single CRDM unit via a connecting rod/spiderassembly or other suitable coupling (details not illustrated). In someillustrative embodiments, a CRDM unit includes a motor driving a leadscrew operatively connected with control rods via a connectingrod/spider assembly, such that motor operation causes linear translationof the assembly including the lead screw, connecting rod, spider, andcontrol rods. Such CRDM units provide fine control of the preciseinsertion of the control rods into the reactor core 10 via the leadscrew, and hence are suitable for “gray rod” operation providing fineincremental reaction control. In some illustrative embodiments, a CRDMunit may comprise a lifting piston that lifts an assembly including theconnecting rod, spider, and control rods out of the reactor core 10, andduring a SCRAM removes the lifting force to allow the control rods tofall into the reactor core 10 by gravity and optional hydraulic pressureforce(s). Such CRDM units are suitably used for “shutdown rod”operation, as part of the reactor safety system. In yet otherillustrative embodiments, the gray rod and shutdown rod functionality isintegrated into a single CRDM unit, for example using a separable ballnut coupling with a lead screw such that the CRDM unit normally providesgray rod functionality but during a SCRAM the ball nut separates torelease the control rods into the reactor core 10. Some furtherillustrative embodiments of CRDM units are set forth in application Ser.No. 12/722,662 titled “Control Rod Drive Mechanism for Nuclear Reactor”filed Mar. 12, 2010 and related application Ser. No. 12/722,696 titled“Control Rod Drive Mechanism For Nuclear Reactor” filed Mar. 12, 2010are both incorporated herein by reference in their entireties. Theseapplications disclose CRDM units providing gray/shutdown rodfunctionality, in which the connection between the motor and the leadscrew is not releasable, but rather a separate latch is provided betweenthe lead screw and the connecting rod in order to effectuate SCRAM. Inthese alternative configurations the lead screw does not SCRAM, butrather only the unlatched connecting rod and control rod SCRAM togethertoward the reactor core while the lead screw remains engaged with themotor.

The diagrammatically illustrated CRDM 24 may include one or more CRDMunits including various combinations of CRDM units of the describedtypes or other CRDM unit configurations providing gray and/or shutdownrod functionality. The illustrative CRDM 24 is an internal CRDM in whichall mechanical and electromagnetomotive components, including the motor,lead screw, connecting rod, and so forth are disposed inside thepressure vessel 12, 14, 16, with only electrical wires, hydraulic lines,or other power or control leads connecting with these components. Inother contemplated embodiments, the CRDM may employ external CRDM unitsin which the motor is mounted outside the pressure vessel, for exampleabove or below.

With continuing reference to FIG. 1, the primary coolant may becirculated naturally, due to natural convection set up by heating due ofthe primary coolant in the vicinity of the operating nuclear reactorcore 10. Additionally or alternatively, the primary coolant circulationmay be driven or assisted by optional reactor coolant pumps 26. Thediagrammatically illustrated coolant pumps 26 are internal pumps havingrotor and stator elements both located inside the pressure vessel 12,14, 16. Alternatively, an external pump can be employed, for examplehaving an external stator and a rotor coupled with the pressure vesselvolume via a suitable conduit or tube, or the circulation pumps may beomitted entirely, as per natural convection reactor embodiments.

The nuclear reactor is further described with continuing reference toFIG. 1 and with further reference to FIGS. 2 and 3. FIG. 2 illustrates aside sectional view of the upper vessel 14 and selected componentstherein, while FIG. 3 shows Section D-D indicated in FIG. 2. As seen inFIG. 1, the illustrative nuclear reactor is an integral nuclear reactor,by which it is meant that a steam generator 30 is integrated inside thepressure vessel 12, 14, 16. In the illustrative example, the pressurevessel 12, 14, 16 is generally cylindrical and defines a cylinder axis A(labeled only in FIG. 2). The steam generator 30 is a straight-tubeonce-through steam generator (OTSG) 30 disposed in the upper vessel 14.The OTSG 30 includes straight tubes 32 arranged vertically in parallelwith the cylinder axis A in an annular “downcomer” volume 34 definedbetween: (i) a hollow central riser 36 disposed coaxially in the upperportion 14 of the generally cylindrical pressure vessel, and (ii) theupper portion 14 of the generally cylindrical pressure vessel. Thehollow central riser 36 defines a central riser flow path 38 inside thecentral riser 36. The OTSG 30 also includes an outer shroud 40surrounding the tubes 32 disposed in the downcomer volume 34, and aninner shroud 42 disposed between the central riser 36 and the tubes 32.(Note that in FIGS. 2 and 3, the OTSG shrouds 40, 42 are shown andlabeled, but the tubes 32 are omitted so as to more clearly show theannular downcomer volume 34 in FIGS. 2 and 3).

The primary coolant flow path in the illustrative reactor is as follows.The central riser 36 has a bottom end proximate to the nuclear reactorcore 10 to receive primary coolant heated by the nuclear reactor core10, and a top end distal from the nuclear reactor core 10. Primarycoolant heated by the nuclear reactor core 10 flows upward through thecentral riser flow path 38 inside the central riser 36. At the top ofthe central riser 36 the primary coolant flow turns 180° (that is, fromflowing generally upward to flowing generally downward). The primarycoolant enters the tubes 32 of the OTSG 30 and flows downward throughthe tubes 32. The primary coolant is discharged from the lower ends ofthe tubes 32 into a primary outlet plenum 44, which passes the primarycoolant flow back to the reactor inlet annulus 18 and back to thereactor core 10.

With continuing reference to FIGS. 1-3 and with further reference toFIG. 4, the outer and inner shrouds 40, 42 of the OTSG 30 define a fluidflow volume of the OTSG 30 between the shrouds 40, 42. This fluid flowvolume surrounds the tubes 32, and has a feedwater inlet 50 and a steamoutlet 52. Note that although a single inlet 50 and single outlet 52 areillustrated, in other embodiments there may be multiple inlets and/ormultiple outlets, to provide redundancy and/or improved radial symmetryin the plane transverse to the axis A. Fluid (e.g., feedwater) isinjected into the fluid flow volume at the feedwater inlet 50 and isdischarged from the fluid flow volume (e.g., as steam) at the steamoutlet 52. While in the fluid flow volume, the fluid flows outside thetubes 32 of the OTSG 30 in a generally upward direction generallyopposite flow of primary coolant inside the tubes 32.

With continuing reference to FIGS. 1-3 and with further reference toFIG. 4, in the operating state of the illustrative PWR, feedwaterinjected into the fluid flow volume of the OTSG 30 at the feedwaterinlet 50 is converted to steam by heat emanating from primary coolantflowing inside the tubes 32 of the OTSG 30, and the steam is dischargedfrom the fluid flow volume at a steam outlet 52. This isdiagrammatically shown in FIG. 4, which shows portions of threeillustrative tubes 32 carrying downward primary coolant flow(F_(primary)). The fluid flow volume of the OTSG 30 is diagrammaticallyshown in FIG. 4 by indication of portions of the outer and inner shrouds40, 42 that define the fluid flow volume of the OTSG 30. To facilitatecorrelation with FIGS. 1-3, the axial direction corresponding to theaxis A of the generally cylindrical pressure vessel is also indicated inFIG. 4. The fluid flowing in the fluid flow volume of the OTSG 30 issometimes referred to herein as “secondary” coolant, and the generallyupward “counter” flow of the secondary coolant in the fluid flow volumeof the OTSG 30 is indicated as secondary coolant flow (F_(secondary)) indiagrammatic FIG. 4. During the upward flow, heat emanating from theprimary coolant flow F_(primary) transfers to the secondary coolant flowF_(secondary), causing the secondary coolant to heat until it isconverted to secondary coolant flow having the form of steam flow(S_(secondary)). (The steam flow S_(secondary) is also diagrammaticallyindicated in FIG. 4 by using dotted arrows). Although not illustrated,the steam flow Ssecondary exiting the steam outlet 52 suitably serves asworking steam that flows to and operates a turbine or othersteam-operated device.

In the illustrative embodiment, the fluid flow volume of the OTSG 30 isdefined by the outer and inner shrouds 40, 42 that are separate from thecentral riser 36 and the upper portion 14 of the pressure vessel.Advantageously, this enables the OTSG 30 to be constructed as a unitincluding the tubes 32 and surrounding shrouds 40, 42, and to then beinstalled as a unit in the upper portion 14 of the pressure vessel.However, it is also contemplated for the inner shroud to be embodied asan outer surface of the central riser 36, and/or for the outer shroud tobe embodied as an inner surface of the upper portion 14 of the pressurevessel.

In embodiments which include the outer shroud 40 which is separate fromthe upper pressure vessel portion 14 (as in the illustrativeembodiment), an annular space between the outer shroud 40 and thepressure vessel 14 may optionally be employed for a useful purpose. Inthe illustrative example, the annular space between the outer shroud 40and the pressure vessel 14 defines a feedwater annulus 60 between anouter shroud 40 of the OTSG 30 and the pressure vessel (upper portion14) buffers feedwater injected into the fluid flow volume at thefeedwater inlet 50. Similarly, a steam annulus 62 between the outershroud 40 of the OTSG 30 and the pressure vessel (upper portion 14)buffers steam discharged from the fluid flow volume at the steam outlet52.

In some embodiments, the feedwater annulus and the steam annulus haveequal inner diameters and equal outer diameters. In such embodiments theouter shroud and the relevant pressure vessel portion have constantdiameters over the axial length of the annuluses. In the illustrativeembodiment, however, the feedwater annulus 60 has a larger outerdiameter than the steam annulus 62. This is obtained by increasing thediameter of the upper pressure vessel portion 14 surrounding thefeedwater annulus 60 as compared with the diameter of the upper pressurevessel portion 14 surrounding the steam annulus 62. In the illustrativeembodiment the diameter of the outer shroud 40 remains constant over theaxial length of the annuluses 60, 62. This configuration allows a largerlocal inventory of water so that the time available for steam generatorboil-off is relatively longer in the event of a loss-of-feedwater (LOFW)accident.

With reference to FIGS. 1 and 2, as already mentioned the flow circuitfor the primary coolant includes a 180° flow reversal as the primarycoolant discharges from the central riser flow path 38 inside thecentral riser 36 and flows into the top ends of the tubes 32 of the OTSG30. Optionally, a flow diverter 70 is provided to facilitate this flowreversal. The illustrative flow diverter 70 is disposed in the generallycylindrical pressure vessel 14 and has a flow diverting surface 72facing the top end of the central riser that is sloped or (asillustrated) curved to redirect primary coolant discharged from the topend of the central riser 36 toward inlets of the tubes 32 of the OTSG30. The flow diverter 70 is spaced apart from the top of the centralriser 36 by a primary inlet plenum 74.

As previously mentioned, the illustrative nuclear reactor is apressurized water reactor (PWR) in which the primary coolant issub-cooled and maintained under positive pressure. In some embodiments,the pressurization of the primary coolant is provided by an externalpressurizer. However, in the illustrative embodiment the pressurizationof the primary coolant is provided by an internal pressurizer. In thisconfiguration, the flow diverter 72 also serves as a part of the dividerplate 70 spaced apart from the top end of the central riser 36 by theaforementioned primary inlet plenum 74. The generally cylindricalpressure vessel 12, 14, 16 (and, more precisely, the upper pressurevessel portion 14) includes a sealing top portion 78 cooperating withthe divider plate 70 to define an integral pressurizer volume 80 that isseparated by the divider plate 70 from the remaining interior volume ofthe generally cylindrical pressure vessel 12, 14, 16. In the operatingstate of the PWR, the integral pressurizer volume 80 contains fluid(saturated primary coolant liquid and steam) at a temperature that isgreater than the temperature of the primary coolant disposed in theremaining interior volume of the generally cylindrical pressure vessel12, 14, 16. In this embodiment, the saturation temperature is maintainedby pressurizer heaters 82 (shown only in FIG. 1), while pressurizerspray nozzles 84 provide a mechanism for reducing the pressure bycondensing some of the steam vapor in volume 80. The divider plate 70suitably includes openings (not shown) providing hydraulic fluidcommunication between the integral pressurizer volume 80 and theremaining interior volume of the generally cylindrical pressure vessel12, 14, 16. This hydraulic fluid communication establishes the pressurelevel in the remaining interior volume of the generally cylindricalpressure vessel 12, 14, 16. Since there is a temperature differenceacross divider plate 70 between the pressurizer volume 80 and primaryinlet plenum 74, the remaining primary fluid in the interior volume ofthe generally cylindrical pressure vessel 12, 14, 16 is maintained atsub-cooled liquid conditions at a temperature approximately 11° C. (20°F.) less than the saturation temperature in pressurizer volume 80. Thislevel of sub-cooled liquid prevents the primary fluid in reactor core 10from experiencing saturated bulk boiling which has a significantlyhigher vapor volume fraction than sub-cooled nucleate boiling typicallypresent in pressurized water nuclear reactor cores. This prevention ofbulk boiling in a PWR core is made possible by the pressurizer (80, 82,84, 78, and 70) and is beneficial for the integrity of the nuclearreactor fuel rods by minimizing the probability of departure fromnucleate boiling (DNB) which increases the fuel pellet and fuel claddingtemperatures.

Having set forth an illustrative integral PWR as an illustrative examplein FIGS. 1-4, some further additional aspects and variants are set forthnext.

In one illustrative quantitative example, the reactor core 10 in theoperating state operates at 425 MW thermal. The hot reactor coolantwater flows in a circuit, called the hot leg, which includes the spaceabove the core flowing around the CRDM's 24. The hot leg extends up thecentral riser 36 to the inlet plenum 74, wherein the reactor coolantsubsequently enters into the tubes 32 of the straight-tube OTSG 30 viathe central riser flow path 38. The straight-tube OTSG 30 encircles thecentral riser 36 and includes the annular array of steam generator tubes32 disposed in the annulus between the central riser 36 and the outershroud 40 of the OTSG 30. An advantage of this configuration is that thecentral riser 36 is a high pressure component separating the highpressure reactor primary coolant at 1900 psia (in this illustrativequantitative embodiment) from the lower pressure secondary coolant whichin this example is at 825 psia. The use of an internal pressure part viathe central riser 36 yields a compact and efficient design since theprimary pressure boundary is internal to the steam generator 30 andserves the dual use as a riser defining the flow path 38 for the hotleg. One design consideration is that there is differential thermalexpansion between central riser 36, the tubes 32, and the upper vessel14. The differential expansion is further complicated by the feedwaterannulus 60 containing feedwater at a substantially lower temperaturethan the steam in the steam annulus 62, which results in a range oftemperatures in the upper vessel 14, causing additional thermal stress.

One approach for mitigating the effect of these differential stresses isto balance the stresses over the operational and non-operational rangeof conditions of the steam generator. In one illustrative example, thetubes 32 are made of an austenitic nickel-chromium-based alloy, such asInconel™ 690, and the tubes 32 are secured in a support made of steel.The support includes an upper tubesheet 90 and a lower tubesheet 92(diagrammatically indicated in FIG. 2) In general, the austeniticnickel-chromium-based alloy will have a higher coefficient of thermalexpansion than the steel. The balancing of the stresses over theoperational and non-operational range of conditions is suitablyaccomplished by pre-stressing the Inconel™ 690 steam generator tubes 32by expanding the tubes 32 into mating holes of the upper and lowertubesheets 90, 92. This expansion draws the tubes into tension via thePoisson effect. In general, the concept is that in the operating stateof the nuclear reactor the primary coolant flowing in the tubes 32 ofthe OTSG 30 is at a relatively high temperature, for example atemperature of at least 500° C., and the tubes 32 of the OTSG 30 aredesigned to be under axial compression in this operating state at hightemperature. On the other hand, the tubes 32 of the OTSG 30 are designedto be under axial tension in a non-operating state of the nuclearreactor in which the tubes 32 are at a substantially lower temperaturesuch as room temperature, for example suitably quantified as atemperature of less than 100° C. The balancing of the stresses over theoperational and non-operational range is achieved by pre-stressing thetubes 32 to be in axial tension at room temperature (e.g., at less than100° C. in some embodiments), so that the differential thermal expansionbetween the Inconel™ 690 steam generator tubes 32 and the steel of thecentral riser 36 and vessel 14 causes the tubes to transition from axialtension to axial compression as the temperature is raised to theoperating temperature, e.g. at least 500° C. in some embodiments. Thesedifferential thermal stresses among components 14, 32, and 36 are set upby common connection of the components at the tubesheet supports 90, 92is also optionally reduced by having the feedwater nozzle 50 positionedlow in the pressure vessel leaving a longer steam outlet annulus 62 toblanket the vessel with high temperature outlet steam, and by reducingaxial length of the feedwater annulus 60 by employing a larger radiusfor the feedwater annulus 60.

With brief reference to FIG. 5, a manufacturing sequence to prestressthe tubes 32 to place them into axial tension is further described. Inan operation 100, the tubes 32 are mounted in the tubesheets 90, 92 ofthe OTSG frame or support by expanding the tube ends to secure them tothe tubesheets 90, 92. A consequential operation 102 is that thisimparts axial tension to the tubes 32. In an operation 104, the OTSG 30including the prestressed tubes 32 is installed in the pressure vessel12, 14, 16 to construct the integral PWR of FIGS. 1-4. In an operation106, the integral PWR is started up and brought to its operating statewhich has the effect of raising the temperature the primary coolantflowing in the tubes 32 of the OTSG 30 to an operating temperature of(in the illustrative example) at least 500° C. A consequential operation108 is that this imparts axial compression to the tubes 32 due to therelatively higher coefficient of thermal expansion of the austeniticnickel-chromium-based alloy of the tubes 32 as compared with the steelof the central riser 36 and vessel 14 connected via tubesheets 90, 92.

In some embodiments, in the operating state the OTSG 30 defines anintegral economizer that heats feedwater injected into the fluid flowvolume at the feedwater inlet 50 to a temperature at or below a boilingpoint of the feedwater. In such embodiments, the straight-tube OTSG 30is an integral economizer (IEOTSG) design since the feedwater is heatedby flow outside of the tubes 32. Feedwater enters through the feedwaternozzles 50, distributes throughout the feedwater annulus 60, and entersthe tubes 32 via a gap or other passage (not shown) between the bundleshroud 40 and the lower tubesheet 92. In the operating mode, feedwaterflows outside of the tubes 32 and there is forced convection heattransfer from the primary coolant flow to the feedwater flow followed bysubcooled and saturated boiling to form the steam flow. Once thecritical heat flux is reached at approximately 95% steam quality, thesteam goes through a transition to stable film boiling followed bydryout at 100% steam quality. Thereafter in the tube bundle, the forcedconvection to steam raises the temperature to superheated conditions atwhich the steam exits the steam generator via the steam outlet annulus62 and the steam outlet nozzle 52. The superheated steam does notrequire moisture separators before the steam is delivered to the steamturbine (although it is contemplated to include moisture separators insome embodiments).

Some further aspects of the integral pressurizer are next described. Thepressurizer controls the pressure of the primary coolant via thepressurizer heaters 82 and the pressurizer spray nozzles 84. To increasesystem pressure, the heaters 82 are turned on by a reactor controlsystem (not shown). To decrease pressure, the spray nozzles 84 injectcold leg water from the top of the reactor inlet annulus 18 on thedischarge side of the reactor coolant pumps 26 via a small external line(not shown). The pressurizer volume 80 is formed by a divider plate 70which separates the space between the primary inlet plenum 74 and thepressurizer volume 80. The divider plate 70 optionally also serves as aflow diverter by including a perforated cylinder 124 (FIG. 6, top ofdivider plate not shown) or a cone shaped flow diverter surface 72 (FIG.2) or other curved or slanted surface which aids in the turning of theflow in the primary coolant in the inlet plenum 74 before it enters theupper ends of the tubes 32 of the OTSG 30 setting up downward flowinside the tubes 30. The illustrative pressurizer including thepressurizer volume 80 and pressure control structures 82, 84advantageously is a fully integral pressurizer (that is, is part of thepressure vessel 12, 14, 16) and advantageously has no pass-throughs forexternal CRDM's or other components.

The central riser 36 forms a path 38 for the primary coolant flowleaving the reactor core 10 to reach the primary inlet plenum 74 of thesteam generator 30. In this embodiment there is no horizontal run ofpiping for this purpose. As a result, if the reactor is operated in anatural circulation mode with the reactor coolant pumps 26 turned off(as may occur during a malfunction or loss of electrical power causingthe pumps 26 to stop operating), the hot rising primary coolant is onlyimpeded by the upper internals (e.g., the CRDM's 24). This flowresistance is not large compared to the flow resistance of the core 10and the steam generator tubes 32 because the flow area is relativelylarge. The flow resistance of the central riser 36 is also a relativelysmall percentage of the total because of the large diameter of the path38.

In some existing nuclear steam supply systems, after a loss of coolantaccident (LOCA) steam and non-condensable gases can collect at the highpoints of the reactor coolant pipes, and can inhibit the naturalcirculation loop between the reactor core and the steam generators.Advantageously, the straight-tube OTSG 30 with integral pressurizervolume 80 disclosed herein automatically removes non-condensable gasesfrom the primary coolant circulation loop since there is only one highpoint at the top of the pressurizer volume 80. Buoyancy causes thenon-condensable gases and vapor to go to the top of the pressurizervolume 80, where these gases and vapor do not interfere with the naturalcirculation loop.

Another advantage of the disclosed straight-tube OTSG 30 is that it canoptionally operate in multiple modes to remove decay heat from thereactor core 10. Starting with the normal operating state, if thereactor coolant pumps 26 stop operating, then the primary coolant watercontinues to circulate, albeit now via natural circulation, through thecore 10 and through the steam generator tubes 32. As long as there isfeedwater supplied to the inlet 50 of the steam generator 30, there is alarge tube surface area to remove radioactive fission product heat fromthe core 10. If the primary coolant level falls below the level of theprimary inlet plenum 74 during a LOCA, then the straight-tube OTSG 30can operate as a condenser. In this mode, steam from boiling water inthe reactor core 10 rises to fill the primary inlet plenum 74 and thepressurizer volume 80. The lower temperature water and steam on thesecondary side (that is, in the fluid flow volume defined outside thetubes 32 by the shrouds 40, 42) causes condensation inside the steamgenerator tubes 30. By gravity alone, the condensate flows out of thesteam generator tubes 32 into the primary outlet plenum 44 where it isreturned to the core 10.

In the straight-tube OTSG 30, the primary coolant pressure is inside thetubes 32. The primary coolant is at a substantially higher pressure thanthe secondary coolant flowing through the fluid flow volume definedoutside the tubes 32 by the shrouds 40, 42. In some embodiments, in theoperating state of the nuclear reactor the primary coolant flowinginside the tubes 32 is at a pressure that is at least twice a pressureof the secondary fluid (feedwater or steam) in the fluid flow volume.This enables the use of a thinner tube wall in tension. In contrast, ifthe primary coolant flows outside the tubes then the tube is incompression and a thicker tube wall is generally required. Some analyseshave indicated that the tube wall in the tension design of the presentOTSG embodiments can be made about one-half as thick as the tube wallthickness required for tubes placed in compression (for comparable tubediameter).

The use of thinner tube walls translates into the OTSG 30 beingsubstantially lighter and including substantially less Inconel™ 690 orother nickel-chromium-based alloy material used for the tubes 32. Theweight saving of the straight-tube OTSG 30 is advantageous for anintegral nuclear reactor. For example, in the illustrative embodiment ofFIGS. 1-3, during refueling the core 10 is accessed by removing thesteam generator 30. This entails disconnecting the OTSG 30 from thelower pressure vessel portion 12 via the mid-flange 16. The lightweightstraight-tube OTSG 30 advantageously reduces the requisite size of thecontainment structure crane used for lifting the steam generator 30 offto the side during refueling.

The straight-tube OTSG 30 also has service and maintenance advantages.Manways are readily provided proximate to the pressurizer volume 80 andthe primary inlet plenum 74 to provide service access. Inspection of thetubes 32 can be performed during a plant outage via the primary inletplenum 74 without removing the steam generator 30 from the pressurevessel. Eddy current inspection thusly performed can reveal tubethinning and tube cracks. If tube plugging is indicated by suchinspection, the steam generator 30 can be removed during the outage andtube plugs can be installed at the lower tubesheet 92 and the uppertubesheet 90. In another approach, both tube inspection and tubeplugging can be done during refueling when the steam generator 30 isplaced off to the side of the reactor. In this case, there is easyaccess from the bottom for tube inspection and plugging.

With reference to FIG. 6, a variant embodiment is described. Thisvariant embodiment includes the IEOTSG 30 with tubes 32 mounted in upperand lower tubesheets 90, 92. In this variant embodiment, a modifiedupper pressure vessel portion 14′ differs from the upper pressure vesselportion 14 in that it does not have a larger diameter to provide afeedwater annulus with larger outer diameter as compared with the steamannulus. Rather, a feedwater annulus 60′ connected with the feedwaterinlet 50 in the variant embodiment of FIG. 6 is of the same outerdiameter as the steam annulus 62 that is connected with the steam outlet52. The modified upper pressure vessel portion 14′ also differs from theupper pressure vessel portion 14 in that it does not include theintegral sealing top portion 78. Rather, a separate sealing top portion78′ is provided which is secured to the modified upper pressure vesselportion 14′ by an upper flange 120. Still further, the variantembodiment also does not include an integral pressurizer volume or thediverter plate 70. Rather, the sealing top portion 78′ defines amodified primary inlet plenum 74′ (but does not define a pressurizervolume), and the sealing top portion 78′ includes a curved surface 122that cooperates with cylinder openings 124 at the top of the centralriser 36 to perform the primary coolant flow diversion functionality.

As the pressurizer volume 80 of the embodiment of FIGS. 1-3 is omittedin the variant embodiment of FIG. 6, primary coolant pressurization forthe embodiment of FIG. 6 is suitably provided by self-pressurization. Inthis approach, steam vapor from the reactor core collects at the top ofthe steam generator vessel, that is, in the primary inlet plenum 74′.The compressibility of the vapor filled dome volume 74′ regulates theprimary coolant pressure. To increase power, the feedwater flow into thefeedwater inlet 50 is increased which increases the boiling lengths inthe tubes 32. The reactor core 10 follows the load demand by increasingpower via a negative moderator coefficient of reactivity due to thereduction in core inlet temperature from the steam generator 30. Thecore outlet temperature is maintained at a near constant temperatureregulated by the pressure and saturation temperature of the steam domevolume 74′. Accordingly, for an increase in power, the temperature riseacross the reactor core 10 increases while the reactor flow rate remainsconstant as determined by the reactor coolant pumps 26. Decreasing poweremploys analogous processes.

As another illustrative variation (not shown), the tubes of the OTSG canbe placed in different locations within the pressure vessel. In theillustrative embodiments of FIGS. 1-3 and 6, the OTSG 30 including tubes32 is disposed entirely in the downcomer volume 34. More generally,however, tubes may be disposed in the downcomer volume (as illustrated),or in the central riser flow path 38 inside the central riser 36, or inboth volumes 34, 36.

As other illustrative variations, it has already been noted that theseparate inner shroud 42 may instead be embodied as an outer surface ofthe central riser 36, and/or for the separate outer shroud 40 mayinstead be embodied as an inner surface of the upper portion 14 of thepressure vessel. Additionally, it is contemplated to integrate the lowertubesheet 92 with the mid-flange 16.

The preferred embodiments have been illustrated and described.Obviously, modifications and alterations will occur to others uponreading and understanding the preceding detailed description. It isintended that the invention be construed as including all suchmodifications and alterations insofar as they come within the scope ofthe appended claims or the equivalents thereof.

We claim:
 1. A nuclear reactor comprising: a generally cylindricalpressure vessel defining a cylinder axis; a nuclear reactor coredisposed in the generally cylindrical pressure vessel; a central riserdisposed coaxially inside the generally cylindrical pressure vessel, thecentral riser being hollow and having a bottom end proximate to thenuclear reactor core to receive primary coolant heated by the nuclearreactor core, the central riser having a top end distal from the nuclearreactor core; and a once-through steam generator (OTSG) comprising tubesarranged parallel with the cylinder axis in an annular volume definedbetween the central riser and the generally cylindrical pressure vessel,primary coolant discharged from the top end of the central riser flowinginside the tubes toward the nuclear reactor core, the OTSG furtherincluding a fluid flow volume having a feedwater inlet and a steamoutlet wherein fluid injected into the fluid flow volume at thefeedwater inlet and discharged from the fluid flow volume at the steamoutlet flows outside the tubes in a direction generally opposite flow ofprimary coolant inside the tubes; wherein the nuclear reactor has anoperating state in which fluid comprising feedwater injected into thefluid flow volume at the feedwater inlet is converted by heat transferfrom primary coolant flowing inside the tubes into steam that isdischarged from the fluid flow volume at the steam outlet; and whereinthe tubes of the OTSG are secured in a support including a pair oftubesheets made of steel and the tubes are supported at their ends bythe tubesheets, wherein the tubes comprise a material having a highercoefficient of thermal expansion than steel and are in axial tension ina non-operating state of the nuclear reactor in which the tubes of theOTSG are at a temperature of less than 100° C. and are in axialcompression in the operating state.
 2. The nuclear reactor as set forthin claim 1, wherein: in the operating state of the nuclear reactor theprimary coolant flowing in the tubes of the OTSG is at a temperature ofat least 500° C.
 3. The nuclear reactor as set forth in claim 2, whereinthe tubes of the OTSG are made of an austenitic nickel-chromium-basedalloy and the tubes are secured in the tubesheets which are attached tothe central riser and the pressure vessel, wherein the central riser andpressure vessel are made of a steel, and wherein the austeniticnickel-chromium-based alloy has a higher coefficient of thermalexpansion than the steel.
 4. The nuclear reactor as set forth in claim1, wherein: the tubesheets are attached to the central riser and thepressure vessel wherein the central riser and the pressure vessel aremade of steel; and ends of the tubes are expanded to secure to thetubesheets whereby the tubes are under axial tension due to the Poissoneffect.
 5. An apparatus comprising: a pressurized water nuclear reactor(PWR) including a pressure vessel, a nuclear reactor core disposed inthe pressure vessel, and a vertically oriented hollow central riserdisposed above the nuclear reactor core inside the pressure vessel; anda once-through steam generator (OTSG) disposed in the pressure vessel ofthe PWR, the OTSG including vertical tubes having a higher coefficientof thermal expansion than steel arranged in an annular volume defined bythe central riser and the pressure vessel and secured in a support madeof steel, the OTSG further including a fluid flow volume surrounding thevertical tubes; wherein the PWR has an operating state in which primarycoolant at a temperature of at least 500° C. flows in the tubes of theOTSG and in which feedwater injected into the fluid flow volume at afeedwater inlet is converted to steam by heat emanating from the primarycoolant flowing inside the tubes of the OTSG, and the steam isdischarged from the fluid flow volume at a steam outlet; and wherein thetubes of the OTSG are in axial tension in a non-operating state of thePWR in which the tubes of the OTSG are at a temperature of less than100° C. and are in axial compression in the operating state.
 6. Theapparatus as set forth in claim 5, further comprising: a flow diverterdisposed in the generally cylindrical pressure vessel above the centralriser, the flow diverter having a flow-diverting surface facing a top ofthe central riser that is at least one of sloped and curved to redirectprimary coolant discharged from the top of the central riser towardinlets of the tubes of the OTSG.
 7. The apparatus as set forth in claim6, wherein the flow diverter divides the pressure vessel into an upperintegral pressurizer volume and a remaining lower interior volume, andin the operating state the upper integral pressurizer volume containsfluid at a temperature greater than a temperature of the primary coolantdisposed in the remaining lower interior volume of the pressure vessel.8. The apparatus as set forth in claim 5, wherein the pressure vessel isdivided into an upper integral pressurizer volume and a remaining lowerinterior volume, and in the operating state the upper integralpressurizer volume contains fluid at a temperature greater than atemperature of the primary coolant disposed in the remaining lowerinterior volume of the pressure vessel.
 9. The apparatus as set forth inclaim 8, further comprising: neutron-absorbing control rods; and acontrol rod drive mechanism (CRDM) configured to controllably insert andwithdraw the control rods into and out of the nuclear reactor core;wherein no portion of the CRDM is disposed in or passes though theintegral pressurizer volume.
 10. The apparatus as set forth in claim 5,wherein: in the operating state of the nuclear reactor the primarycoolant flowing inside the tubes is at a higher pressure than the fluidin the fluid flow volume.
 11. The apparatus as set forth in claim 5,wherein the tubes of the OTSG are prestressed to place the tubes inaxial tension in the non-operating state of the PWR in which the tubesof the OTSG are at a temperature of less than 100° C. by operationsincluding: mounting the tubes in the tubesheets by expanding the tubeends to secure them to the tubesheets, whereby axial tension is impartedto the tubes.
 12. An apparatus comprising: a pressurized water nuclearreactor (PWR) including a pressure vessel, a nuclear reactor coredisposed in the pressure vessel, and a vertically oriented hollowcentral riser disposed above the nuclear reactor core inside thepressure vessel; a once-through steam generator (OTSG) disposed in thepressure vessel of the PWR, the OTSG including vertical tubes having ahigher coefficient of thermal expansion than steel arranged in anannular volume defined by the central riser and the pressure vessel andsecured in a support made of steel, the OTSG further including a fluidflow volume surrounding the vertical tubes; neutron-absorbing controlrods; and a control rod drive mechanism (CRDM) configured tocontrollably insert and withdraw the control rods into and out of thenuclear reactor core; wherein the PWR has an operating state in whichprimary coolant at a temperature of at least 500° C. flows in the tubesof the OTSG and in which feedwater injected into the fluid flow volumeat a feedwater inlet is converted to steam by heat emanating from theprimary coolant flowing inside the tubes of the OTSG, and the steam isdischarged from the fluid flow volume at a steam outlet; wherein thetubes of the OTSG are in axial tension in a non-operating state of thePWR in which the tubes of the OTSG are at a temperature of less than100° C. and are in axial compression in the operating state; wherein thepressure vessel is divided into an upper integral pressurizer volume anda remaining lower interior volume, and in the operating state the upperintegral pressurizer volume contains fluid at a temperature greater thana temperature of the primary coolant disposed in the remaining lowerinterior volume of the pressure vessel; and wherein no portion of theCRDM is disposed in or passes though the integral pressurizer volume.13. The apparatus as set forth in claim 12, further comprising: a flowdiverter disposed in the generally cylindrical pressure vessel above thecentral riser, the flow diverter having a flow-diverting surface facinga top of the central riser that is at least one of sloped and curved toredirect primary coolant discharged from the top of the central risertoward inlets of the tubes of the OTSG.
 14. The apparatus as set forthin claim 13, wherein the flow diverter divides the pressure vessel intoan upper integral pressurizer volume and a remaining lower interiorvolume, and in the operating state the upper integral pressurizer volumecontains fluid at a temperature greater than a temperature of theprimary coolant disposed in the remaining lower interior volume of thepressure vessel.